Features of Divertor Plasmas in W7-AS
Y. Feng, P.Grigull, F.Sardei, K.McCormick,
J.Kisslinger
 Island divertor vs tokamak divertor
-basic differences and similarities
 Transport features of attached plasmas
 Recycling, neutral screening & core-fueling
 Impurity screening of the edge islands
 Detachment and detachment stability
Island divertor concept for low-shear stellarators
W7-AS
W7-X
 An intermediate low-order island chain between confinement core and
plasma-surface Interaction region, screening the core from direct
penetration of recycling neutrals and sputtered Impurities
 Island divertor experiments in W7-AS from 2000 to 2002
Divertor vs limiter in W7-AS
 Easier density control even in the presence of strong
NBI-sources
- Significant improvement of recycling condition and
particle pumping
 High Density H-mode (HDH) maintainable quasistationary
- concomitant dense, cold plasma in the edge islands
 Strong reduction of diverted energy flux onto targets
via impurity line radiation
- Existence of stable partial detachment in certain
geometry and plasma parameter ranges
- Intensive radiation outside confinement region, no serious
degradation of global energy content
Island divertor vs tokamak divertor
Tokamak
Poloidal-field divertor
single-null
double-null
Stellarator
island divertor
from single- to multi-null
W7-AS: 8,9,10
Standard=9
W7-X: 4,5,6
Standard=5
1D SOL transport model
core
X-P.
-V
+V
dx 
a / R  B p / B  0.1 tokama k
 
ds 
W7 - AS
ri ' a / R  0.001
( introduced only for simple analysis)
target
d 
dT  d 
dT 
5/ 2
    n
  0
    eT 
dx 
dx  dx 
dx 
nd c sd Td  qII


d
d 
d
2

mnV II  p     D mnV II
dx
dx 
dx

mnV II
  D
2


y / 2

Stellarator specific
Flow damping & momentum loss
nV (EMC3)
Schematic expression
V =0
radial x
V =0
-V
+
V =0
+V
poloidal y
radial
target
target

toroidal
-
An extended 2P-model with -transport
(no CX-momentum loss, no volume energy losses)
7/ 2
up
T
T
7/ 2
d
7qII Lc 7 n


T  Td
2 up
2 e 2 e 


nd Td c sd   qII
D  1 nVII / nd c sd 
pup  2 pd 1  f m ; f m 

dx
2
2c sd  
y / 2 
Large  -> standard 2-point model
Extended-2P-model results
-transport strongly damps downstream evolution
No high-recycling regime
EMC3/EIRENE
Langmuir probes on targets
Island neutral screening & particle re-fueling
Experiment
Island
screening
low T
Low n
EMC3/EIRENE
• Recycling provides the main fueling source for the core
• Island screening -> flattening of fueling rate
• In the flattening range, gas-puff increases only the SOL density, rather than
the core density
• High density core correlates with dense, cold islands
Dense, cold islands shift CX-neutrals
to a low energy band
Energy spectrum of CX-neutrals
hitting the Fe-wall
EMC3/EIRENE
CX
H-atom / eV-1
nes=5×1019 m-3
1×1019
E0 /eV
In/out asymmetry in reducing high-energetic
CX-neutral on wall
(E0 > threshold energy for Fe-sputtering)
nes=1×1019 m-3
nes=5×1019 m-3
Sensitivity of Fe-yield to nes, ne(r) and D
EMC3-EIRENE
→A high density in the edge islands is the most efficient parameter for
reducing the wall-sputtering yield.
Impurity retention under high-density condition
force balance
VZ II  Vi II  2.2
 Vi II 
1D radial continuity:
VZ II
 Zi
mZ
 Zi
mZ
Z 2 II Ti 
2.2 Z
2

 Zi
m Z nZ
 1  II Ti 
 II nZ Ti
 Zi Ti
m Z nZ
 II nZ  VZ* II  DZ II
 II nZ
nZ
d 
dn 
 nZVZ* II  ( DZ    2 DZ II ) Z    S Z  RZ  S Z 1  RZ 1
dx 
dx 
(Z-independent)

Z
d 
dn
  nIVZ* II  DZ  I
dx 
dx

  0 (source - free)

Solution for target-released impurities:
Core
Island SOL
0
4cm X
 target VZ* II

nIs  nId exp    
dx 


D
Z

X
point


A positive (outwards), large V*ZII reduces the impurity density at separatrix!
Simple analysis contd.
Condition for V*zII>0 :
5 / 2niTiVi II
friction
~
1
5/2
thermal force  iTi  IITi
(KRASHENINNIKOV,
Nucl. Fusion 1991)
1D energy transport for ion:
5 / 2niTiVi II    iTi 5 / 2 2
dTi
dT
  i ni i  qi
dx
dx
Because of the small  (~0.001) in W7-AS, the parallel heat conduction
can be significantly reduced by the perpendicular one. The latter becomes
even dominating under the condition:
ni / Ti5 / 2   i 2 /  i
=> high-n, low-T SOL plasma favorable for SOL impurity retention
Impurity flow reversal
low edge density
thermal force >
high edge density
EMC3/EIRENE
thermal force
friction
+
=
<
+
inwards flow
friction
outwards flow
=
Strong reduction of C density at separatrix
(normalized to total carbon yield in order to isolate transport from production)
thermal-forces draw carbon
to separatrix
EMC3/EIRENE
low nes
low nes
frictional plasma flow flushes
carbon back to targets
high nes
high nes
Sharp transition from thermal-force dominated
to friction-dominated transport
EMC3/EIRENE
ni / Ti5 / 2   i 2 /  i
ni / Ti5 / 2   i 2 /  i
Friction
dominates
Thermal force
dominates
• sharp transition because of high sensitivity of classical heat conductivity to Ti
High density for detachment transition
absence of a high-recycling regime
shift of detachment transition
to a high nes
EMC3/EIRENE with PSOL = 1 MW
Intrinsic carbon
Td=10 eV
Extended 2P-model
0.5
1
2
3
4
5
Abrupt detachment transition observed in experiments under conditions:
abs
PNBI
 1.4 MW , n e  3 1020 m 3 , n es  6 1019 m 3
In good agreement with the code prediction
Detachment stability depends on island geometry
Experiments
Experimental
results
change of Lc
Lc para. targ-X-p. dist. /m
change of x
P/MW
E /kJ
X radial targ-X-p. dist. /cm
#56846
#56848
stable
#56843
#56847
stable
PNBI
Prad
time (density ramp)
time
Stable partial detachment
Experimental finding: a) stable detachment requires large islands with large 
b) stable detachment always partial
Power load on target
EMC3/EIRENE
EMC3/EIRENE
C-radiation
H-ionization
(inboard side)
(divertor region)
Thermography
hot spot
Marfe-like phenomena
small islands or
~ density limit
field-line pitch
(independent of island geometry)
•Complete detachment
•Weak neutral screening EMC3/
•Unstable
•Unstable (exp.)
EIRENE
•Strong degradation of e
Increase
density
• Unstable, intensive radiation zone appearing at the inboard midplane
observed by a CCD camera when plasma approaches density limit.
Impact of radiation location on neutral screening
sensitivity of neutral screening to configuration, nes and Psol
Psol=1 MW
0.8 MW
core
recyc
/
NBI
EMC3/EIRENE
nes 1013 cm-3
Divertor radiation  cold recycling zone  less efficient for neutral screening
‘less efficient’ means: 1) higher recyc into core (smaller X)
2) more sensitive to change of nes or Psol (radiation location)
Summary
Plasma:
• The -to-‖ transport ratio can be changed from <1 to >1,
depending on divertor configurations and plasma parameters.
• Flow-damping -> no high recycling regime.
Recycling neutrals:
• Weak neutral screening -> strong edge-core coupling
- recycling neutrals are the main fueling source for the core
- recycling and refueling process nonlinear -> instabilities
For example, the abrupt change of edge plasma state observed in density-ramp
experiments and the geometry-related detachment instability.
Impurity:
 Dense, cold islands – favorable for reducing influx of intrinsic
impurities
- Reduction of impurity sputtering yield from CX-neutrals
- Frictional plasma flow flushes Impurities
Summary contd.
Detachment:
Detachment transition:
• High densities needed for detachment
• Abrupt change in radiation level and location
Stability (depending on island geometry)
• stable detachment is always partial and needs sufficientlylarge islands
-inboard-side radiation -> warm recycling region -> good
neutral screening
• small islands -> divertor radiation -> loss of neutral screening
-> unstable
Marfes (unstable)
• appear always on the inboard side, inside LCFS whenever a
plasma approaches density limit, independent of configuration.
Principle: Divertor vs limiter
q
target

core
Limiter
impurity
radial
separatrix
q

core
Divertor
impurity
 An intermediate SOL exists between confinement core and plasma-surface
Interaction region, screening the core from direct penetration of
recycling neutrals and sputtered Impurities.